The Sensitivity Analysis Of Reactivity Feedback Of Thermal-Hydraulics
The simulation of the nuclear power station is a significant way to design, assess safety and improve economic benefit of nuclear power station for nuclear scientists. There is an important phenomenon in pressure water reactor that the reactivity of nuclear reaction is related to the thermal-hydraulic parameters of reactor core. The reactivity changes due to the change of thermal-bydraulic parameters, and moreover affects the results of 3-D neutronics simulations. The temperature of fuel and moderator influence the macro cross section by influencing micro cross section (σ) of related nuclide, and the density of moderator, void fraction and the concentration of dissolved boron influence the macro cross section by changing the density (n) of nuclide. It is evident that the results of thermal-hydraulic calculations are important. Nowadays, becanse of the complexity of flow and thermal conduction, some correlations in thermal-hydraulic simulation are experienced or semi-experienced, the correlations are sometimes used outside the range of their validity and are approximately implemented in the code, which bring some uncertainties in the result. In this paper, qinshan-I reactor coolant system is simulated by the coupled codes of REMARK neutron kinetics and THEARe thermal-hydraulics. The correlations of heat transfer coefficients usually used in single fluid phase heat transfer simulation are presented. Comparing the results of the different correlations used in simulation, the semitivity of reactive feedback of thermal-hydraulic is analyzed.
Sensitivity Reactivity Feedback Correlations Simulation
Ma Tingwei Cao Xinrong
Department of Nuclear Science and Technology, Harbin Engineering University 150001, China
国际会议
ISSNP2008、CSEPC、ISOFIC2008(第二届21世纪和谐核电系统国际会议、第四届电厂控制中认知系统工程方法国际会议暨第三届未来核电厂仪表与控制国际会议)
哈尔滨
英文
314-321
2008-09-08（万方平台首次上网日期，不代表论文的发表时间）